ML061800269
| ML061800269 | |
| Person / Time | |
|---|---|
| Site: | Surry  (DPR-032, DPR-037) |
| Issue date: | 06/29/2006 |
| From: | Stephen Monarque Plant Licensing Branch III-2 |
| To: | |
| Gratton C, NRR, DORL, 415-1055 | |
| Shared Package | |
| ML061710263 | List: |
| References | |
| TAC MD1236, TAC MD1237 | |
| Download: ML061800269 (8) | |
Text
3.
This renewed license shall be deemed to contain and Is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 7.0.32 of 10 CFR Part 70; and Is subject to all.applicable provisions of the Act and the..
rules, regulations, and orders of the Commission now or hereafter in effect; and Is subject to the additional conditions specified below:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2546 rriegawatts (thermal).
B.
Technical Specifications The Technical Specifications.contalned In Appendix A, as revised through Amendment No.. 248 are hereby Incorporated in the renewed license.
The licensee shall operate the facility in accordance with the Technical Specificalions.
C.
Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D.
Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.
E.
Deleted by Amendment 65 F.
Deleted by Amendment 71 G.
Deleted by Amendment 227 H.
Deleted by Amendment 227 Fire Protection The licensee shall implemeht and maintain In effect the provisions of the approved fire protection program as described In the Updated Final Safety Analysis Report and as approved In the SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April 27, 1982, November 18, 1982, January 17,1984, February 25, 1988, and Renewed License No. DPR-32 Amendment No, 248 E.
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
. This renewed license shall be deemed to contain and is subject to. the conditions.
specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of IOCFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and Is subject to the additional conditions specified below:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not In excess of 2546 megawatts (thermal).
B.
Technical Specifications I*
The Technical Speciftcations contained in Appendix A, as revised through Amendment No.2-247 are hereby Incorporated in this renewed license.
The licensee shall operate the facility In accordance with the Technical Specifications.
C.
Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D.
- Records The licensee shall keep facility operating records in accordance with the "requirements of the Technical Specifications.
E.
Deleted by Amendment 54 F.
Deleted by Amendment 59 and Amendment 65 G.
Deleted by Amendment 227 H.
Deleted by Amendment 227 Renewed License No. DPR-.37 Amendment No. 247
TS 3.1-9 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units 1 and 2, respectively. The most limiting value of RTNDT (228.40F) occurs at the 1/4-T, 00 azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RTNjrD; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (B greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTDr at the end of 28.8 EFPY and 29.4 EFPY for Units I and 2, respectively (as well as adjustments for location of the pressure sensing instrument).
Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EPPY for Units 1 and 2, respectively, prior to a scheduled refueling outage.
AmendmentNos. 248/247
TS 3.1-10 Allowable pressure-temperature relationships far various heatup and cooldown rates are calculated using methods derived from Appmnx G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanirs (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one halfT is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Sectiom Mfl as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-dnctile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temimeature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress inAensity factor, KI, for the combined thermal and pressure stresses at any time during heatp or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR is obtained from the reference fracture tougtmess curve, defined in Appendix G to the ASME Code. The KW, curve is given by the equution:
K]R=26.78+1223exp[0.0145(T-RTtIf+ 160)1
/
(1) where KIR is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperatme RTND. Thus, the governing equation for the heatup-cooldown analysis is defmed in Apendix G of the ASME Code as follows:
CKIM+ Kit <KIR (2) where, KIM is the stress intensity factor caused by membrance (pressure) stress.
AmendmentNos. 248/247
TS 3.1-11 Kit is the stress intensity factor caused by the thermal gradients Ka is provided by the code as a function of temperature relative to the RTNryr of the material.
C = 2.0 for level A andB service limits, and C = 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, Ki, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60TF per hour. The cooldown limit curves of Figure 3.1-2 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside. wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cooldown limit curves are valid for cooldown rates up to 100*F/br.
The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively. The adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (AIRVSP) documented in the most recent revision to BAW-1543 and reactor vessel neutran fluence data obtained from plant-specific analyses.
AmendmentNos. 248/247
TS 3.1-23a (3)
During the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, maintain a bubble in the pressurizer with a maximum narrow range level of 33%,
(Jr (4) Maintain two Power Operated Relief Valves (PORV) OPERABLE with a lift setting of < 390 psig and verify each PORV block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or (5) The RCS shall be vented through one open PORV or an equivalent size opening as specified below:
(a) with the RCS vented through an unlocked open vent path, verify the path is open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or (b) with the RCS vented through a locked open vent path verify the path is open at least once per 31 days.
- 2.
Themiuirements of Specification 3.1.G.1.c.(4) may be modified as follows:
a One PORV may be inoperable in INIERMEDIATE SHUTDOWN with the RCS average temperature > 200°F but < 350°F for a period not to exceed 7 days. If the inoperable PORV is not restored to OPERABLE status within 7 days, then completely depressurize the RCS and vent through one open FORV or an equivalent size opening within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b One PORV may be inoperable in COLD SHUTDOWN or REFUELING SHUTDOWN with the reactor vessel head bolted for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inoperable PORV is not restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> then completely depressurize the RCS and vent through one open P0RV or an equivalent size opening within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
AmendamntNos. 248/247
Figure 3.1-1 Surry Units I and 2 Reactor Coolant System Heatup Limitations 2500.00 0)CL 0
4D 1500.00 1000.00 0
2.'
-1500.O0 0.0.
Material Property Basis Urrifflng Material: Suny Unit I friterinecriate to Lower Shell Ciro Weld UmItIng ART Vskies for Surry I at 28.8 EFPY. 1/4-T. 228AF 3/4-T, 189.5 F leak Test Und acceptawe Operation I A 1 1,10 IA=spftble Heatup Rate3 oooo rag (Ffi,4 20 4ý 0
50 100 150 200 250 300 350 400 Figure 3.1-1:
Indicated Cold Leg Temperature (Deg. F)
Surry Units I and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 6 0WFjli-) Applicable for the first 28.8 EFPY for Surry Unit 1 and the first 29.4 EFPYfr Surry Unit 2 AmendmentNos. 248/247
Figure 3.1-2 Surry Units I and 2 Reactor Coolant System Cooldown Limitations Material Property Basis Lining Materiat Sum/y Unit I Intemiedlate to Lower Shell Circ Weld Willing ART Values for Su/y I at 28.8 EFPY: 114-T, 228.4F 314-T, 189.5 F 2500.00 2000.00 1500.00 C-0 11000.00 V
4Si to 5WO III IIV I I VT Unacceptable rmagon I v Cacklom ftes Operaum FRIO 40 IM
_L - A-f 0.00 0
50 100 150 200 250 300 350 400 Figuse 3.1-2:
Indicated Cold Leg Temperature (Deg. F)
Suny Units I and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 0-/hr) Applicable for the first 28.8 EFPY for Surry Unit 1 and the first 29.4 EFPY for Surry Unit 2 Amendment Nos. 248 / 24 7